Comparison of innovative concepts to help in the decision of the UST_2  fusion experiment.  Spherical Tokamak  and IEC. (Part 3)    

  

 

Abstract :  Basic features are analysed for several fusion concepts, particularly the ones that have implications in the engineering cost and difficulty. Stellarators and tokamaks excluded. Cost of coils and maintenance, Beta,  possibility of advanced fuels, current drive, energy confinement times, particle handling and reactor studies are analysed. The study is divided in 3 Parts, two devices per page. The comparison of the basic features and recent experimental results should help in the decision of a new fusion device, UST_2. If there is not a device  more convincing than a stellarator, UST_2 may be a stellarator.

 

 

* Spherical tokamak  and IEC  (Devices analysed in Part 3)

     

See Part 2 and 3 for :

*  Dipole and FRC

Spherical tokamak  and IEC

 Stellarator and tokamak not studied here.

 

 

   One and a half years ago the analysis of different devices was omitted due to lack of knowledge [100].  The objective is to clarify the alternatives to design a small device for fusion research in engineering. In [100] the result was to build UST_1 stellarator which finally was designed and built as an optimised modular stellarator with external resistive coils on a circular toroid, plaster frame for grooves and copper vacuum vessel. UST_1 has obtained good magnetic surfaces and some initial plasmas.

 

   Important additional knowledge has been acquired from then, many spreadsheets are of immediate use, a flexible and powerful JAVA code has been developed to simulate, calculate and optimise configurations of coils, many creative and risky techniques, successful and unsuccessful, have been experimented. Building a new stellarator similar to UST_1 would be a matter of one or two intensive months of work and about 300 of cost. However such a possibility lacks of sense.  

 

   Some new key engineering features should be included in the new design. Additionally the physics design should be improved, however it might depend on the help of experts in plasma physics.

 

   UST_2 does not need to be a stellarator. Any other fusion device with better perspectives as a reactor could be chosen.

 

   In principle  UST_2 is not intended to be constructed without some funding and/or interest from foundations, university, company, sponsor, etc.

 

  The next compilation will be expanded in the future if new or more accurate information is found. 

  Maybe a table will be completed when all the devices had the same basic data. It will help to have a global vision.

 

 

 


 

  

   No-one can be an expert in all fusion devices so some inaccuracies surely will appear in the next comparisons of devices. If you know about one of the devices as a fusion reactor and would like to correct some mistakes, add recent data or an opinion, please send further experimental data supported by reasons. Scaling laws, based even in very reduced experimental results only to have an order of magnitud, are extremely important in concepts barely studied.

 

  

 

 

SPHERICAL TOKAMAK

 

References [50] to [59] plus general ones. 

 

Named also Compact Tokamak or Spherical Torus

 

The  experimental data for Spherical Tokamaks is notable. 

 

 

Devices : MAST , START (now Italy) , SPHEX in UK ; NSTX, CDX-U , HIT-II , PEGASUS in USA ; TS-3, TST in Japan , ETE in Brazil , Globus-M in Russia ; Rotamak-ST, Australia. 

 

Advantages : (Comparison with conventional tokamaks). Smaller size for the same plasma, high Beta so lower cost of coils, easy maintenance if copper coils, no major disruptions, almost total CD from bootstrap so possible CW. 

 

Disadvantages (Comparison with conventional tokamaks). Important difficulties in divertor, central post and power handling in general.  Cost of the non-inductive start-up. Regular confinement of Alphas.

 

 

Stability :  Better than large Ap tokamaks at the same beta.

 

Alpha heating regime : Unknown but should be similar to conventional tokamaks.

 

Energy confinement time and transport:

MAST: H98pby2=1.5 #4580(2002) [53]. HIPB98(y,2) =ave.-of-several=0.9 [52]. Similar to classical tokamaks and less than advanced tokamaks [3]. Trapped particles 75%.

NTSX: H98PBy,2=ave-of-several=0.8 [54]

 

Beta ave. :  MAST : Beta=11.5%, BetaN =4.7 (2002) [53]. Beta = 7% [52] . START has the record.

NSTX  : Beta max=35%, BetaN~6.4 (not simultaneously) [55] but stability is compromised (internal pressure driven kink modes). 

 

Advanced fuels :  Difficult. Stable Beta (~10-20%), and confinement like classical tokamaks.  

One study of reactor [56] : R=8m , r=6m , 6GW, Ave. power density=0.3MW/m3, Aux. Heating=0.8GW , Beta=80%, Ip=100MA, Bt=2.7T, Boostrap Ip = ?% . Not very competitive.

 

Self-heated and CW/pulsed :  YES . CW

 

Current drive  : Low proportion or null. Around 20% achieved.

 

Particle handling : Critical issue for reactor design. Only pebble design and/or FW change every 2 years possible. Central rod and divertor issues.

 

ST with LiWalls  : See several papers in [57].  Supported by Leonid Zakharov and other researchers. The performace seems highly improved so this concept must be further studied (find new data and possible critical weakness). LiWalls for stellarators is cited.

 

Example of reactor :  [50] [51] :  R=3.5 , r=2.5 , Ip=31MA , n=1e20m-3, Beta ave =60% , Bo=? , CD = 50MW , HIPB98=1.5 , Neutron wall loading=4MW/m2 ,  bootstrap I = 90% , alphas lost =5% . Copper central rod receives 88MW neutron load and withstands 6 years, rod shield thickness=12cm,     FW and blanket changed every 2 years, 21% of heat is of low T (70-200C), pebble bed, TF power supplies = 32MA x 8V = 256MW (high recirculating power, expensive, 16 exact independent units)

 

Critical issues :   - Cooper central rod and its power supplies~recirculating power  - High neutron and divertor loads. - Cost of start-up CD (perhaps Coaxial Helicity Injection (CHI) together with traditional ones)

 

 

 

 

 

 

 

IEC

 

Inertial electrostatic confinement

 

 

References [30] to [39] plus general ones. 

 

 

Devices : The devices in University of Illinois (3 devices) , Los Alamos National Laboratory Marshall Space Flight Center University of Wisconsin , Greatbatch, Ltd.,  some in Japan [30] and some amateur.

 

The  experimental data for IEC  is still very incomplete

 

Only gridded IEC devices studied here. Virtual cathode devices still suffer from instabilities of the virtual cathode [31]

 

Advantages : It is far simpler than magnetic fusion. Easy access for maintenance. Small devices allow to test the behaviour of larger ones. Ash does not accumulate. Very suited for advanced fuels. Able for direct conversion of energetic particles. Simple maintenance of blanket in case of D-T. No disruptions.

 

Disadvantages : Grid cannot be used in reactor regime. Fusion ash leaves the confinement without transferring energy (so large recirculating power). Relatively small central high density volume so low power at acceptable sizes. Research not focussed, in general, in the production of energy [30].

 

 

Stability :  No applicable. 

 

Alpha heating regime : Not clear. Alphas escape from confinement but relevant collisions may occur at high density. Could degrade focus or reduce efficiency of direct conversion [1]

 

Energy confinement time and transport:

In IEC devices the main energy losses may be caused by transport of electrons or ions (direct loss of fusion products) and to bremsstrahlung radiation [31]

 

 

Beta :  Not applicable

 

Advanced fuels : Very adequate for D-3He and D-D 

 

Self-heated and CW/pulsed :  NO . Externally-heated (acceleration of ions), CW

 

Current drive  : NO.  Really it is energy to accelerate ions up to the required energy.

 

Particle handling : Almost impossible on the grid for reactor powers. External wall does not seem a major issue.

 

Data for scaling laws :  At 120kV ,  70mA, 20cm diam. grid cathode, year 2004, neutron production =~ 1e8 neutrons/s for D-D , 2.45MeV neutrons. [34] [33] [32] . Doubling the size of the cathode grid resulted in 20% increase of neutron production [34]. The high density nucleus supposed to be 2cm diameter for the 10cm grid. Cathode power supply = 8400W .  7.25MeV per neutron (sup. no catalized D-D, so no energy from 3He reaction), total = 0.1mW.

 

Example of reactor :  Not found

 

Critical issues :   - Grid destruction at reactor power (neutrons in D-T, neutrons in D-3He, Helium bombardment, particle heat load, radiation.  

- In not gridded devices, how to mantain a powerful virtual cathode.

- Recirculating power for the accelerator-grid supply.

 

 

 

 

 

Beam-driven Inertial Fusion

 

 Impossible to research in Inertial fusion with poor means due to short pulses, extreme pulsed power and laser-beam technologies.

 

 

 

 

 

 Reasons and insights to chose a suitable

device for UST_2

 

     A decision will be taken some time after the end of the study of the devices.

 

  

  

 

 

 References

 

GENERAL

[1] "Comments on Innovative Confinement Concepts
and Burning Plasma Science" E. B. Hooper

[2] "Fusion research : Experiments " T. J. Dolan , Pergamon Press.

[3] "Fusion Energy Science Opportunities in Emerging Concepts" Los Alamos National Laboratory Report -- LA-UR-99-5052

[4] "EXCITING OPPORTUNITIES TO ADVANCE FUSION ENERGY IN MAGNETIC CONFINEMENT CONCEPTS"  TONY S. TAYLOR et al.

[5] "MFE Concept Integration and Performance Measures" Magnetic Fusion Concept Working Group ,  M.C. Zarnstorff, D. Gates, E.B. Hooper, et al.

 

 

SPHERICAL TOKAMAK

[50] "The Spherical Tokamak Fusion Power Plant" H R Wilson 1), G Voss 1), J-W Ahn 2), R J Akers 1), L Appel 1) et al. (2002)

[51] "Development of the Spherical Tokamak Power Plant" G.M. Voss, A.Bond1, J.B. Hicks, H.R. Wilson (~2000)

[52] "Energy and Particle Confinement in MAST" M. Valovi , H. Meyer, R. Akers,

[53] "Results from the MAST Spherical Tokamak" A.Sykes, J-W Ahn1, R. J. Akers, E. Arends2, K.Axon,

[54] "Confinement Scaling and Transport Properties in NSTX Plasmas" S.M. Kaye1, M. G. Bell1, R. E. Bell1, E. D. Fredrickson1,

[55] "β-Limiting MHD instabilities in improved-performance NSTX spherical torus plasmas" J.E. Menard et al 2003 Nucl. Fusion 43 330-340
 

[56] "Core Plasma Characteristics of a Spherical Tokamak D-3He Fusion Reactor"  Shi Bingren,  Plasma Science & Technology, Vo1.7, No.2, Apr. 2005

[57] LiWalls concept (Several papers)

"LTX the Lithium Tokamak eXperiment" R. Majeski, R. Kaita, L. Zakharov, P. Efthimion,

 

"Getting serious about Fusion" Leonid E. Zakharov, Fusion Theory Colloquium July 21, 2006, UKAEA Fusion, Culham, UK.
 

"Ignited spherical tokamaks and plasma regimes with LiWalls" L.E. Zakharova,, et al.  Fusion Engineering and Design 72 (2004) 149168

 

"Ignited Spherical Tokamaks as a Reactor Development Facility" Leonid Zakharov et al.

 

 

 

IEC

[30] "Progress in Inertial-Electrostatic Confinement Fusion"
John F Santarius, Gerald L Kulcinski, and Robert P Ashley

[31] Brief Overview of Inertial-Electrostatic-Confinement Fusion
John F. Santarius

[32] "Recent progress in Steady State Fusion using D-3He "  R.P. Ashley, G.L. Kulcinski,  J.F Santarius, et al. ,  Fus. science and Technology , 44 , 2003

[33] http://fti.neep.wisc.edu/iec/RecordsandResults.htm

[34] "Optimizing Neutron Production Rates from D-D Fusion in an IEC device" A.L. Wehmeyer et al. , poster

 

 

 

[100]

"First approaches for the design and construction of an ultra small torus to experiment with fusion technologies" Vicente M. Queral. See "List of all R&D"

 

 

 

 

 

 

 

 

 

 

 


 

Date of publication 26-02-2007. Continuous addition of data